ML20096G715
| ML20096G715 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/21/1992 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20096G674 | List: |
| References | |
| NUDOCS 9205260212 | |
| Download: ML20096G715 (11) | |
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ATTACHMENT I to JPN 92-021 gjRRENT TECHNICAL SPECIFICATIONS TO BE CHANGED REGARDING REMOVAL OF REACTOR VESSEL HEAD SPRAY PORTION OF RHR PIPING AND ASSOCIATED VALVES i
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1 New York Power Authority JAMES A. FFFZRATRICK NUCi_ EAR POWER PLANT Docket No. 50-333 9205260212 920521 PDR ADOCK 05000333 P PDR
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l JAFHPP TABLE 4.2-1 ti1HIHuH . TEST, AND.CEIBRATIQt!JPEQUStiCLf9R PCIS Instrwnent_ChanncLJ3) InstIDment.,JunsLionaLTest Caliba tinn_ frequency. Innttsment,_ Check t 4 )
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- 1) Reactor Illgh Pressure (1) 'Once/3 months None (Shutdown Cooling Permissivel
- 2) Reactor Low-Low-Low Water; Level . (1)(5) (15) Once/ day
- 3) Main Steam liigh Temp. l (1)(5) (15) 'Once/ day
- 4) Main. Steam Illgh Flow (1)(5) (15) da - may
- 5) Main Steam Low Pressure (1)(5) (15) Once/ day *
- 6) Heactor Water Cleanup High Temp. (1) Once/3 months .None
- 7) Condenrer Low Vacuum (1)(5) (15) Once/ day LoglC Sys L.fA__fmantiO901_ Test _D )E _ _ __,._ _ . _ _ . . _ _ _ _ . _ _ _ .._.___ _ fIequency______ _ ._. _ _ _ _ _ . _ _ _ _ _
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- 1) Main Steam Line Isolation valves Once/6 months Main Steam Line Drain Valves Reactor Water Sample Valves
- 2) RHR - Isolation valve Control Once/6 months lea r
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- 3) Reactor Water Cleanup Isolation Once/6 months
- 4) Drywell Isolation Valves Once/6 months Tip Withdrawal Atmospheric Control Valves
- 5) Standby Gas Treatment S3 stem Once/6 months Reactor Building Isolation NOTE: See listing of notes following Table 4.2-6 for the notes referred to herein.
Ame nfime r.t No . ((, , 136 -
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1 ATTACHMENT 11 to JPN-92 021 PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING REMOVAL OF REACTOR VESSEL HEAD SPRAY PORTION OF RHR PIPhlG AND ASSOCIATED VALVES JPTS-92-016 O,
-- 4 Ncw York Power Authority JAMES A. FITZPA'AICK NUCLEAR POWER PLANT xet No. 50 333 l
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TABLE 4.2-1 .
MINIMUM TEST /sND CAllBRATION FREQUENCY FOR PCIS Inctrument Channel (8) instrument Functional Test Calibration Frequency Instrument Check (4)
Reactor High Pressure _ (?) Once/3 months None (Shutdown Cooling Permissive)
M Reactor Low-Low-Low Water Level (1)(5) (15) Once/ day
- 3) Main Steam High Temp. (1)(5) (15) Once/ day
- 4) Main Steam High Flow (1)(5) (15) Once/ day
- 5) Main Steam Low Pressure (1)(5) (15) Once/ day
- 6) Reactor Water Cleanup High Temp. (1) Once/3 months Nor'e
- 7) Condenser Low Vacuum (1)(5) (15) Once/ day Logic System Functional Test (7) (9) Frequency
- 1) Main Stean. Line iso!ation val"es Once/6 months Main Steam Une Drain Valves Reactor Water Sample Valves
- 2) RHR -Isolation Valve Contre' Once/6 months Shutdown Cooling Valves
- 3) Reactor Water Cleanup Isolation Once/6 months
- 4) Drywellisolation Valves Once/6 months TIP Withdrawal A!mospheric Control Valves
- 5) Standby Gas Treatment System Once/6 months Reactor Building isolation NOTE: See listing of notes fc! lowing Table 4.2-6 for the notes referred to herein.
Amendment No. 37,89,136, 78
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ATTACHMENT lli to JPN-92-021 3 I
SAFETY EVALUATION FOR j
PROPOSED TECHillCAL SPECIFICATION CHANGES REGARDING REMOVAL Of REACTOR VESSEL HEAD SPRAY PORTION OF RHR
,'IPING AND ASSOCIATED VALVES JPTS-92-016 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333
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Attachment lll to JPN 92-021
- SAFETiEVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGE REMOVAL OF REACTOR VESSEL HEAD SPRAY PORTION OF RHR PIPING AND ASSOCIATED VALVES (JPTS-92-016)
Page 1 of 6
. I, . DESCRIPTION OF THE PROPOSED CHANGES The proposed change to the James A. FitzPatrick Technical Specifications revises Table 4.21 entitled, "Minirnum Test Calibration Frequen:y for PCIS" on page 78. Part of item (2) j
" Head Spray' is deleted from the table, j l
ll. PURPOSE OF THE PROPOSED CHANGES MODIFICATION TO REMOVE HEAD SPRAY l
- The putpose of this change is to reflect a plant modification which will deactivate the reactor
. vessel head spray portion of the Residual Heat Removal (RHR) System. This modification involves the removal of portions of the spray pipe and hangers from the flanged elbow attached to the cpray nozzle connection on the reactor vesdai head (the flanged elbow will l be retained) to the west (inboard) side of the missile protectic wall at elevation 336'-0".
Figure 1 providas a schematic of the Reactor Vessel Head Spey System. Valves iOMOV-32,10MOV 33 and check valve 10RHR-29 are located in the portions of pipe that will be removed. The end of the retained piping will be capped. The spray piping running through the drywell penetration thermal sleeve will be cut near both ends of the penetration, and a cap installed on each end of the pipe. This will be the primary containment boundary for this penetration. The existing LLRT connection on the pipe (including va!/es 10RHR-708 and 10RHR-709) located on the spray pipag just outside the drywcll, w'.il be retained for uso during preopertional testing of the new end caps. An ASTM blind flange will bo installed m
= the retained flanged elbow at the reactor vessel head to ensure reactor vessel pressure iintegrity. All retained piping will be seismically analyzed to ensure system integrity. This r.1odification will be completed during the 1992 Refueling Catage.
The Authority analyzed the head spray's function and design and determined that the head spray system may b' deactivated without causing any affects to other systems that perform safetyrrelated functions.
. CONTAINMENT ISOLATION VALVES h
One reason for this modification is that the containment isolation valves (CIVs),10MOV-32 and 10MOV43 recently failed the local leak rate test (LLRT). T hese CIVs automatically isolate the process line which penetrates the primary conta!nment for head spray. Head =
spray is an optional capability of RHR which is not used at FitzPatrick.
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Attachment lit to JPN-92 021 SAFETY EVALUATION Page 2 of 6 MODIFICAT10t! NO. F1-92-091 DEACTIVATION OF RHR HEAD SPRAY SYSTEM i
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l NEW'8UND FLANGE i NEW PIPE CAP' ff SEAL SUPPOST DRYWELL PENETRATION' N" tlEW PIPE !
- RING CAP n
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CONTAINMENT n f80UNDARY/' .Y/
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REACTOR % P $
as-vr p w as I VESSEL
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= m]-3y y 10RHR-708H FT 10RHR-709 R i
-NEW PlPE NEW PIPE- I" CAPS CAPS l l MISSILE SHIELD WALL 2 y,, ',
RHR SYSTEM
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= TO BE RETAINED PlPiNG
= TO BE REMOVED PIPING REACTOR VESSEL HEAD SPRA" SCHEMATIC
Attachment lit to JPN 92-021 SAFETY EVALUATION Page 3 of 6 The Authority determined that a modification to eliminate system maintenance and l surveillance testing of these valves could be done. This determination was based upon 1
schedular constraints estimated high repair costs for the CIVs, and the fact that head spray is an optional capability of RHR which is not used at FitzPatrick.
PERSONNEL RADIATION EXPOSURE Another reason for this mcdification is to reduce personnel radiation exposure during vessel disassembly and reassembly. The flanged section of the head spray piping within the drywell must be unbsted and removed prior to removal of the reactor vessel head and then reinstalled prior to dartup. This section of t. < iping will be permanently removed by this modification reducing personnel radiation exposure in future maintenance outages associated with this removable pipe section.
This modification will also reduce personnel radiation exposure associated with the repair of the containment isolation valves and the Inservice inspection (ISI) of the piping and components in this subsystem.
111. SAFETY IMPLICATION OF THE PROPOSED CHANGES FUNCTION OF HEAD SPRAY
! The head spray pcrtion of the RHR symm supplies water to the vessel steam dome through the head spray nozzle at low reactor pressures. It is intended for use during shutdown cooling to enhance reactor vessel head cooling with the remainder of the vessel metal below the water line. In Reference 1, General Electric (GE) stated that this design feature was installed to reduse outage time based on the assumption that the vessel cooldovin and head removal would be critical path activities. At Fitzpatrick, head spray is not used, and head cooldown is not on the outege cr;!ical path. The head spray is an optional capability and credit is not taken for it in the accident analysis.
The moddication dros not degrade the capability of RHR to meet its safety objective. As stated in Section 4.8.3 of the Updated FSAR, the safety objective reads, "The objective of l the RHR System is to restore and maintain the coolant inventory in the reactor vessel so that the core is adequately cooled after a LOCA." The head spray is not used to restore or maintain reactor vessel water level after a LOCA. The low pressure coolant injection (LPCI) mode of RHR performs this function. Head spray may be used to augment the RHR system in the shutdown cooling mode to provide normal shutdown cooling. However, as documented in Roference 2, head spray is not required, and the design of RHR is adequate l~ for cooldown without head spray.
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6 Attachment ill to JPN 92-021 SAFETY EVALUATION Page 4 of 6 The head spray system is not described in the basis for any Technical Specification. This subsystem is not required to perform any safety-related functions.
Head Spray is mentioned in the Emergency Operating Procedurt (EOPb10, " Primary Containment Flooding.' in this EOP, Head Spray is listed as one of ten available sources of water that may be used to maintain containment water level between 85 ft. and 105 ft. This procedure is only used if there is a need to flood the drywell under emergency conditions.
The Head Spray system will be removed from EOP 10 when this modification is completed.
FUNCTION OF PRIMARY CONTAINMENT ISOLATION SYSTEM The Primary Containment Isolation System (PCIS) provides timely protectica against the consequences of accidents involving the release of radioactive materials from the fuel and
- Reactor Coolant Pressure Boundary. The PCIS initiates automatic idation of appropriate process lines which penetrate the primary containment whenever monitored variables exceed preselected operational limits.
- This modification removes the containment isolation valves 10MOV 32 and 10MOV-33 from the head spray portion of the RHR system. Since the bead spray containment isolation valves will be removed, the logic system functional test for these valves will not be necessary.
The PCIS for other CIVs will not be affected by this change. The logic functional test for other portions of PCIS will continue at the current frequency. Control room panels 9-3 and 9-4 will be revised due to the removal of indicators, switches, and indication lights.
lI SIMILAR CHANGES AT OTHER BWRs Functionally similar modifications have been performed at BWR plants including Hatch Units 1 and 2 and Brunswick Units 1 and 2. This modification was evaluated by General Electric
- - (GE) for Hatch in Reference 1. The GE evaluation concluded that, " Removal of the head spray capability of the RHR system was found to have no significant impact on plant safety or operations while providing substantial benefits for plant capacity factor and personnel
- radiation exposure reduction. Consequently, removal of the function is recommended.'
In May of 1990, Carolina Power & Ught Company submitted a request for two license l - amendments regarding the emoval of RHR head spray flow transmitter for Brunswick Unit 1 and 2. In January of 1991, the NRC imod license amendments 151 and 181 regarding the removai of residual heat removal head spray flow transmitter fcr Brunswick Units 1 and 2 L
(Reference 3). This arrendment request is similar to the license amendnients the NRC approved for Brunswick Units 1 and 2 to raflect the deactivation of head spray function.
Attachment til to JPN-92-021 SAFETY EVALUATION Page 5 of 6 IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with this proposed amendment would not involve a significant hazards consideration, as defined in 10 CFR 50.92, since the proposed changes would not:
- 1. lavolve a significant increase in the probability of an accident or consequence previously evaluated.
This chaage will not increase the possibility of an accident or malfunction of safety-related structures, systems or components as evaluated previously in the ;
FSAR. There are no safety related functions associated with the operation of '
head spray Head spray is an optional capability and credit is not taken for it in the accident analysis.
- 2. create the possibility of a new or different kind of accident from those previously evaluated.
The proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated because there are no new interfaces with safety-related equipmer-1, systems or structures. No new systems have been introduced which by their failure or malfunction could create a new or different accident.
The change deletes 'no hgic system functional test for the head spray containment isola'.<on valv as (CIVs) that will be removed as part of the plant modification. The Primary Containment isolation System (PCIS) for all other CIVs will not be affected by this change. The logic functional test for other portions of PCIS will continue at the current frequency.
- 3. involve a significant reduction in the margin of safety as defined in the basis for Technical Specifications.
The change will not reduce the margin of safety as defined !n the Technical Specification.
The head spray system is not described in the basis for an3 Tmhnical Specification. This subsystem is not required to perform any safety-related functions. Head spray is an optional caoability and credit is not taken for it in the accident analysis.
1 Attachment ill to JPN 92-021 l SAFETY EVALUATION !
Page 6 of 6 V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not impact tne ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes impact the environment.
VI. CONCLUSION These changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:
- a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety prsviously evaluated in the safety analysis report;
- b. will not increase ihe possibility for an accident or malfun , tion of a different type from any evaluated previously in the safety analysis report;
- c. will not reduce the margin of safety as defined in the basis for any technical specification; and
- d. Involva no significant hazards consideration, as de:ined in 10 CFR 50.92.-
Vll. REFERENCES
- 1. General Electric Company Report No. MDE Memo 109-1284 (DRF #E00-00157),
" Evaluations to Justify Head Spray Removal *, dated Dec3mber 1984.
- 2. General Electric letter, John Cihl to Alan Ettlinger, dated May 13,1992, "NYPA/FitzPatrick DBD Development Program RHR DBD/Rev B - Reactor Vessel Head Spray Sub-System."
- 3. NRC letter, Ngoc Le to Lynn Eury, dated January 9,1991, " Issuance of Amendment 15I and Amendment 181 Regarding Removal RHR Head Spray Flow Transmitter -
Brunswick Steam Electric Plant, Units 1 and 2."
- 4. James A. FitzPatrick Nuclear Power Plant Technical Specifications, Section 4.2,
" Surveillance Requirements Instrumentation."
- 5. ' James A. FitzPatrick
- uclear Power Plant Updated Final Safety Analysis Report, Sections 4.8, " Residual Heat Removal System" and 7.3, " Primary Containment and Reactor Vessel isolation Control System."
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